压水堆燃料棒三维热-力学性能的精细模拟

压水堆燃料棒工作在复杂的辐照、热和力学环境中,对其性能进行定量评估涉及多种复杂的物理现象。目前常用的燃料性能分析程序一般对结构采用简化的轴对称假设,对辐照肿胀、辐照蠕变和高温蠕变等物理现象以及辐照-热-力等物理场之间的耦合考虑并不充分。基于ABAQUS有限元求解框架,开发了压水堆燃料棒三维热-力学性能的模拟程序,利用程序对压水堆燃料棒进行了稳态分析,以及升功率和反应性引入事故两种瞬态分析。结果表明:辐照引起燃料致密化和肿胀对燃料温度变化有重要影响;芯块应变增加主要是由裂变产物肿胀引起的;芯块几何结构导致包壳应力集中发生在芯块间的交界面处;燃料棒功率的急剧变化会加快芯块表面破裂的进程;反应性引入事故会导致芯块从内部开始破裂,并会引发芯块-包壳的接触。     

弥散型燃料板辐照肿胀行为的有限元分析

采用颗粒复合材料的细观力学研究思路,针对弥散型燃料板,选择一种代表性单元作为研究对象,用与裂变密度和燃耗相对应的、不断增大的裂变压力模拟燃料颗粒裂变对铝基体所产生的力学贡献,对研究对象进行了热力耦合分析.考察了裂变压力、温度应变和燃料颗粒沿厚度非均布对颗粒肿胀的影响,并对弥散型燃料板基体的米塞斯应力分布情况进行了分析.研究结果表明,颗粒百分比含量越高,颗粒肿胀的速度相对越快;随着裂变压力增大,颗粒肿胀的速率急剧增大,且颗粒不再是圆形;在正常工作和选取的参数条件下,温度应变对总变形的影响不大;当裂变压力不大时,非均匀分布对颗粒肿胀几乎没有影响,随着裂变压力的增大,非均匀分布的颗粒肿胀明显高于均布情况,而且越来越明显;基体的应力非球对称分布.     

U_3Si_2Al弥散型燃料的辐照肿胀研究

介绍了U3Si2 Al弥散型燃料的辐照肿胀机理。将弥散型燃料的芯体视为连续基体中的微型燃料元件 ,应用裂变气体的行为机理描述燃料相中的气泡形成过程。研究结果表明 :燃料相的肿胀引起燃料颗粒和金属基体之间的力学相互作用 ,金属基体能抑制燃料颗粒的辐照肿胀。在一定辐照条件下 ,本模型对燃料元件辐照肿胀的预测值与测量值相符     

U3Si2—Al弥散型燃料的辐照肿胀研究

介绍了U3Si2-Al弥散型燃料的辐照肿胀机理。将弥散型燃料的芯体视为连续基体中的微型燃料元件,应用裂变气体的行为机理描述燃料相中的气泡形成过程。研究结果表明：燃料相的肿胀引起燃料颗粒和金属基体之间的力学相互作用,金属基体能抑制燃料颗粒的辐照肿胀。在一定辐照条件下,本模型对燃料元件辐照肿胀的预测值与测量值相符。     

热蠕变对UMo/Zr单片式燃料板起泡行为的影响

针对含有气腔的UMo/Zr单片式燃料板,考虑包壳材料的热蠕变效应,将包壳的变形与气腔压力相耦合,发展了一种对燃料板宏观起泡行为进行数值模拟的方法。基于所建立的模拟方法,计算分析了包壳热蠕变和气腔内裂变气体原子数对起泡行为的影响。研究发现,在考虑包壳热蠕变时,若局部开裂区域内的裂变气体原子数为4.0×1017,以鼓泡高度0.1 mm作为起泡阈值的判断标准,所预测出的阈值温度比不考虑热蠕变时低100℃;若局部开裂区内的裂变气体原子数由2.5×1017增加至4.0×1017,则燃料板的起泡阈值温度将可能降低40℃,通过降低包壳材料的热蠕变率可以有效提高燃料板的抗鼓泡能力。      

弥散核燃料等效辐照肿胀计算模拟

本文将弥散核燃料芯体看作一种特殊的颗粒复合材料,利用细观计算力学的方法,假设燃料颗粒在芯体中周期性分布,建立了对芯体等效辐照肿胀进行计算模拟的有限元模型。考虑颗粒的辐照肿胀和基体材料的辐照硬化效应,分别建立了燃料颗粒和基体材料的应力更新算法,编制了用户材料子程序,在Abaqus软件中实现了芯体等效辐照肿胀的有限元模拟。计算分析了颗粒大小和体积含量对芯体等效辐照肿胀的影响,并得到了等效辐照肿胀的拟合公式。研究结果表明,影响芯体等效辐照肿胀的主要因素是颗粒的辐照肿胀和体积含量。     

压水堆燃料辐照性能多物理耦合并行分析程序开发

核燃料元件是反应堆的核心部件,其性能影响反应堆的安全性与经济性,利用燃料元件性能分析程序开展燃料堆内稳态辐照性能分析对于燃料设计及安全评价具有重要意义。通过开发燃料温度分布、变形计算、裂变气体释放及内压等模型,结合燃料元件热工-力学多物理耦合计算分析耦合方案,基于先进并行计算方法构建了高性能并行化燃料性能分析程序Athena。利用典型商用压水堆核电站数据及同类程序计算结果进行了程序初步验证,结果表明Athena程序计算结果合理可靠。通过定义堆芯功率及热工水力边界条件,程序能够并行开展压水堆全堆芯燃料辐照性能分析,提高燃料辐照性能分析效率,是数值反应堆原型系统（CVR1.0）的重要组成。     

在板型核燃料元件服役行为模拟中引入离散辐照数据的方法

核燃料元件的辐照工况条件一般只能以离散数据的形式给出,如何将其引入元件堆内热力耦合行为的计算模拟中成为一个关键性的问题。本文针对板型弥散核燃料元件,建立了一系列的方法,可将大量的离散辐照数据处理为所需要的数据格式,从文件中精确搜索到所需要的数据点和辐照数据,并插值到元件热力耦合分析单元的积分点。通过编制FORTRAN程序,将离散辐照数据成功地引入定义等效芯体和包壳材料热-力学本构关系的用户材料子程序、定义等效芯体产热率的子程序,从而将离散辐照工况数据引入了元件热力耦合服役行为的有限元计算。通过输出元件堆内行为有限元计算中所实际使用的工况数据,验证了离散辐照数据引入方法的正确性和高效性。     

辐照蠕变对锆合金包壳管吸氢所致多场耦合行为的影响

本文考虑辐照效应,改进了锆合金包壳管内部的氢原子扩散-氢化物析出-热-力耦合行为的微分控制方程。根据多物理场等效积分弱形式和所建立的耦合计算方法,在FEPG软件平台编制文件,生成多场耦合计算的有限元程序,并对程序进行了验证。计算分析了辐照蠕变对锆合金包壳管堆内吸氢所致多场耦合行为演化的影响,结果表明:辐照蠕变导致包壳管内产生应力松弛,促使Mises应力显著降低,同时导致静水应力由负值转变为正值,进而影响氢原子的扩散;与不考虑辐照蠕变的结果进行对比,发现辐照蠕变会增大燃料芯块与包壳管局部接触区域的负的静水应力的绝对值及向外的静水应力梯度,导致接触区域内的氢原子浓度减小,接触区域周围的氢原子浓度增大。     

TRISO燃料颗粒三维多物理场耦合计算模型开发

三向同性燃料(TRISO)颗粒是高温气冷堆弥散型燃料和全陶瓷微密封(FCM)耐事故燃料芯块的裂变区。为研究TRISO燃料颗粒在辐照环境中的复杂行为,基于COMSOL有限元软件开发了TRISO燃料颗粒的三维多物理场耦合性能分析模型。通过采用随辐照条件变化的材料物性参数和行为模型,可模拟燃料颗粒在稳态运行和事故工况下复杂的堆内热-力学行为,以及CO气体产生和裂变气体释放、裂变产物扩散等重要物理过程,还可以计算燃料颗粒的失效概率。基于COMSOL开发三维分析模型的计算结果与美国BISON程序对TRISO燃料颗粒的计算结果相比同样符合较好,说明了所开发模型的合理性。     

Simulation of pellet-cladding thermomechanical interaction and fission gas release

This paper summarizes the present status of a computer code that describes some of the main phenomena occurring in a nuclear fuel rod throughout its life. Temperature distribution, thermal expansion, elastic and plastic strains, creep, mechanical interaction between pellet and cladding, fission gas release, gas mixing, swelling, and densification are modeled. The modular structure of the code allows for the incorporation of models to simulate different phenomena and material properties. Collapsible rods can be also simulated.The code is bidimensional, assumes cylindrical symmetry for the rod and uses the finite element method to integrate the differential equations. The stress–strain and heat conduction problems are nonlinear due to plasticity and to the temperature dependence of the thermal conductivity. The fission gas inventory is calculated with a diffusion model, assuming spherical grains and using a one-dimensional finite element scheme. Pressure increase, swelling and densification are coupled with the stress field.Good results are obtained for the simulation of the irradiation tests of the first argentine prototypes of MOX fuels, where the bamboo effect is clearly observed, and of the FUMEX series for the fuel centerline temperature, the inside rod pressure and the fractional gas release.     

A measurement of the irradiation-induced creep of mixed carbide nuclear fuel

In the BR 2 reacior at Mol, Belgium, a measurement of the irradiation induced creep of mixed carbide nuclear fuel up to high burnup was carried out The dependence upon applied stress and burnup of 95% dense (U, Pu) C was measured within a temperature range between 500 and 720°C and at fission rates between 1.0−1.5 × 10¹⁴ f/cm³ s. The used irradiation device was a Confluent-type capsule that allowed a variation of stress as well as temperature during irradiation. The length changes of the fuel specimen were determined by means of the microwave cavity resonance method. The obtained creep rates are proportional to stress and burnup-independent. The irradiation creep rates are about one order of magnitude below those of mixed oxide fuel. The fission product swelling rate increased with burnup form initially 1.2 to 3.0 vol% per % burnup. At stress changes the fuel showed a transient swelling up to 0.2 vol%. The theoretical background of carbide irradiation creep is briefly discussed.     

高燃耗陶瓷燃料颗粒的参数化建模及强度分析

弥散燃料芯体中的陶瓷燃料颗粒在辐照条件下会形成裂变气孔,燃料颗粒内部气孔间的相互干涉作用及气孔内压的增长致使局部拉应力超过材料强度极限,进而导致燃料颗粒开裂。本文考虑高燃耗燃料颗粒内气孔尺寸和位置分布的非均匀性,实现了颗粒内部的细观结构参数化建模。运用有限元方法计算并分析了气孔尺寸、基体约束压应力、温度和气孔分布方式对颗粒内部最大拉应力的影响,研究了颗粒内开裂危险区的分布规律。结果表明,陶瓷燃料颗粒最大拉应力随气孔尺寸和温度的增加而增大,随基体约束压应力的增加而减小;燃料相的断裂强度减小,开裂危险区面积增大;燃料颗粒从内部多处开裂破坏,而表层处开裂的概率更大。本文为弥散燃料失效研究及优化设计提供了分析方法及数值参考。     

Diametral strain of fast reactor MOX fuel pins with austenitic stainless steel cladding irradiated to high burnup

The C3M irradiation test, which was conducted in the experimental fast reactor, “Joyo”, demonstrated that mixed oxide (MOX) fuel pins with austenitic steel cladding could attain a peak pellet burnup of about 130 GWd/t safely. The test fuel assembly consisted of 61 fuel pins, whose design specifications were similar to those of driver fuel pins of a prototype fast breeder reactor, “Monju”. The irradiated fuel pins exhibited diametral strain due to cladding void swelling and irradiation creep. The cladding irradiation creep strain were due to the pellet-cladding mechanical interaction (PCMI) as well as the internal gas pressure. From the fuel pin ceramographs and ¹³⁷Cs gamma scanning, it was found that the PCMI was associated with the pellet swelling which was enhanced by the rim structure formation or by cesium uranate formation. The PCMI due to cesium uranate, which occurred near the top of the MOX fuel column, significantly affected cladding hoop stress and thermal creep, and the latter effect tended to increase the cumulative damage fraction (CDF) of the cladding though the CDF indicated that the cladding still had some margin to failure due to the creep damage.     

Effect of non-equilibrium fission gas and fuel creep on swelling and release in irradiated carbide fuels

The code UCSWELL was developed to simulate fission gas behavior in carbide fuels. In the present work, one of the limiting assumptions in UCSWELL - that matrix gas bubbles are in equilibrium with gas atom concentration - is removed and non-equilibrium matrix fission gas bubbles are allowed, but with relaxation to equilibrium by means of vacancy diffusion and thermal and radiation-induced creep of the fuel. For a given grain size, the difference in swelling between equilibrium and non-equilibrium with relaxation bubble fission gas treatment increases with decreasing irradiation temperature. At a given temperature, the non-equilibrium effect is more pronounced for larger grain fuel. This is to be expected because the creep rate (and hence the rate at which bubbles grow to an equilibrium size) decreases as temperature decreases and/or as grain size increases. At temperatures, where the creep rate is grain size insensitive, grain size remains important to the equilibrium process in so far as the grain boundary is a source of vacancies to the non-equilibrium bubbles. While the difference in these quantities is at the most on the order of 20% for the steady operating conditions considered, it is anticipated that the non-equilibrium effects become more pronounced during reactor overpower and undercooling transients.     

Thermal Stress Analysis of High-Burnup LWR Fuel Pellet Pulse-Irradiated in Reactivity-Initiated Accident Conditions

For RIA-simulated experiments in the NSRR with high-burnup PWR fuel and BWR fuel, numerical analyses were performed to evaluate the temporal changes of profiles of temperature and thermal stress in pellet induced by pulse power, using the RANNS code. The pre-pulse states of rods were calculated using the fuel performance code FEMAXI-6 along the irradiation histories in commercial reactors and the results were fed to the RANNS analysis as initial conditions of the rod. One-dimensional FEM was applied to the mechanical analysis of the fuel rod, and the calculated cladding permanent strain was compared with the measured value to confirm the validity of the PCMI calculation. The calculated changes in the profiles of temperature and stress in the pellet during an early transient phase were compared with the measured data such as the internal gas pressure rise, cracks and grain structure in the post-test pellet, anddiscussed in terms of PCMI and grain separation. The analyses indicate that the pellet cracking appearances coincided with the calculated tensile stress state and that the compressive thermal stress suppresses the fission gas bubble expansion leading to grain separation.